ML20064F009

Forwards Appl for Amend to Lic#NPF-2:proposed Changes to Tech Specs Will Eliminate Reactor Trip Following Turbine Trip Below 50% Pwr,Revove Part Length Control Rods, & Add Maint Superintendent to Plant Staff

ML20064F009
Person / Time
Site:	Farley
Issue date:	11/15/1978
From:	Clayton F ALABAMA POWER CO.
To:	Schwencer A Office of Nuclear Reactor Regulation
References
NUDOCS 7811270176
Download: ML20064F009 (34)

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cc .e-- s-s v D 5'C'" 4 S .IO-e-g2-z.a .a Te w., m: w F. L CLAYTON. A sewvee P es.*= AlabamaPower tre southem e:ecte system November 15, 1978 DocketNo.50-348[

Director of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D.C. 20555 Attn: Mr. A. Schwencer D

Dear Mr. Schwencer:

RE: Changes to Operating License No.

NPF-2 Technical Specifications Alabama Power Company proposes the attached changes to Joseph M. Farley Nuclear Plant Operating License No. NPF-2 Technical Specifications involving the following items:

~~

1. Technical Specification 4.2.2.2.e.2 and Bases B3/4.2.2, B3/4.2.3 concerning changes in Fxy limits. This change is deemed not to involved a significant hazards conside-ration and is considered a Class III change according to 10 CFR Part 170.

2. Technical Specification Table 3.3-1 and Bases page B 2-7

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concerning elimination of a reactor trip following a tur-

' bine trip below 50% power. This change is deemed not to involve a significant hazards consideration and is con-I sidered a class III change according.to 10 CFR Part 170.

3. Technical Specifications 1.13, 3.1.3'.1, 4.1.3.1.2, Table 3.1-1, 3.1.3.2, 3.1.3.3, 3.1.3.7, 4.1.3.7, 4.2.2.2, 3.2.4, j 3.10.1, 4.10.1.3, 3.10.2, 3.10.3, B3/4.2.1, and 5.3.2 con-I cerning removal of part length control rods. This change I is demmed not to involve a significant hazards considera-tion and is considered a class III change according to 10 CFR Part 170.

4. Technical Specification 6.5.1.2, 6.5.3.1, and Figure 6.2-2 '

concerning addition of a Maintenance Superintendent to the plant staff. This is an administrative change which is considered as a class II change according to 10 CFR Part 170.

7 811270 l~lfo'

Director of Nuclcar Reactor Regulation 1:ovember 13, 1978 Page 2 The Plant Operations Review Committee and the Nuclear Operations Review Board have reviewed the above proposed changes and have determined that the changes do not involve an unreviewed safety question as shown in the attached safey essluations.

The class of each item in this proposed amendment is designated according to 10 CFR Part 170 requirements. A check for $12,300 is enclosed to cover the total amount of fees required.

In accordance with 10 CFR 50.30 (c)(1)(i), three (3) signed originals and thirty-seven (37) additional copies of these proposed changes are enclosed.

If you have any questions, please advise.

Ycurs very truly, b o F. .klayton,Jr.

FLCJr/TNE:bhj Enclosures cc: Mr. R. A. Thomas Mr. G. F. Trowbridge

[ SWO i TO AND SUBSCRIBED BEFORE ME THIS s DAY OF NOVDBER,1978.

b!OTARY PUBLIC V My Cormtission Expires: [- M *[ d

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, . ENCLOSURE RADIAL PEAKING FACTOR (Fxy)

PROPOSED TECHNICAL SPECIFICATION CHANGES (1) Delete: 4.2.2.2.e.2 on-page 3/4 2-7 Replace with:

2. For all unrodded core planes:

1 FxyRTP < 1.68 up to core elevations of 2.4ft.

e FxyR TP 71.75 for core elevations from 2 4ft to 7.2ft.

FxyRTP ][ 1.61 for core elevations above 7.2 ft.

(2) Change: B 3/4.2.2 and B 3/4.2.3 heading to include Fxy on page B 3/4 2-4 With: 3/4.2.2 and 3/4.2.3 HEAT FLUX and NUCLEAR ENTHALPY RISE HOT CHANNEL FACTORS AND RADIAL PEAKING FACTORS - FQ(z),FgH and Fxy (z)

(3) Add: discussion on Fxy peaking factor in the Bases of 3/4.2.3

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, on page B 3/4 2-5 l The radial peaking factor, Fxy (z), is measured periodically

to provide additional assurance that the hot channel factor, Fn (z), remains within its limit. The Fxy (z) limits were determined from expected power control maneuvers over the full range of burnup conditions in the core.

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. SAFETY EVALUATION FOR Fxy - PEAKING FACTOR CHANGE

Background:

~

Recent analyses for Westinghouse plants which undertake reload cycle ,

operations with an Fxy technical specification show the need to revise l Fxy (z) limits upward for subsequent cycles of operation. Farley Unit 1 l is one of the first Westinghouse plant with an Fxy tech spec to reload, '

thus a revision is necessary at this time to avoid exceeding this tech spec during the second and subsequent cycles.

References:

(a) Technical Specifications 4.2.2.2, B 3/4 2-4 (b) WCAP-8385, " Power Distribution Control and Load Following ?rocedures",

I. Sept. 1974.

Bases:

The heat flux hot channel factor Fg (z), which is the primary power distribution parameter in the Technical Specifications for LOCA protection, is determined by the product of the radial Fxy and the axial F (z) peaking factors. The current Farley 1 Tech Spec (reference (a)) requires Fxy be

< l.55 for unrodded planes and 5,1.71 for rodded planes, and that Fg (z)

Ee less than 2.32 x K(z). However, in the event that the Fxy tech s exceeded, continued operation is allowed provided the operating z)is FQ (pec is evaluated to be within its tech spec limit.

Recent analysis by Westinghouse have identified a need to revise Fxy I

limit upward to accomodate expected variations in Fxy for cycle 2 and subsequent cycles. Since reload cores exhibit flatter axial sha z) still indicated by lower F(z), the increase in Fxy will result in Fg (pes, as being within allowable limits.

For the remainder of cycle 1 and cycle 2 of the Farley core, bounding values of the peaking factor FQ (z) x (relative power) were calculated as a function of elevation by assuming various load follow transients on the reactor through insertion and removal of control rod Banks C and D. The l effects of the accompanying variation in axial xenon and power distributions l

were also considered as described in the Reference (b). Both beginning and end of cycle conditions were included in the cycle 2 calculations, and several different histories of operation were assumed in calculating effects

of load follow transients on the axial power distribution. Results of these calculations demonstrate that the Fo (z) tech spec will not be exceeded .

for the remainder of cycle 1 and alt of cycle 2. Further evaluations indicate that the proposed Fxy (z) change will not cause the Fg (z) technical specification limits to be exceeded in projected reload cycles beyond cycle 2. The Fg (z) limit envelope will be reverified for each reload beyond cycle 2 to confirm the projection.

Conclusion:

l Since the present FQ (z) limits will not be changed, there is no de:rease in safety margin associated with the proposed change in Fxy limits.

Therefore, this change does not constitute an unreviewed safety question as defined by 10CFR50.59.

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POWER DISTRIBUTION LIMITS

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SURVEILLANC REQUIREMENTS (Continued)

C RTP

2. When the F is less than or equal to the F limit for.

the appropf5 ate measured core' plane, adpitidnal power dgributiopmapsshallbetakenandF compared to F and F at least once per 31 EFPDYY xy xy e'. The F limits for RATED THERMAL POWER within speci,fic core planefYshall be:

e 1. F < l .71 for all core planes containing bank "D" i cdntrol rods and/or any part length rods, and m .-

RTP , , ce

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f. The F limits of e,'above, are not applicable in the fol-lowinfcore plane regions as measured in percent of core height from the bottom of the fuel:

1. Lower core region from 0 to 15%, inclusive.

2. Upper core region from 85 to 100%, inclusive.

3. Grid plane regions at 17.8 + 2%, 32.1 + 2%, 46.4 + 2%, - -

60.6 + 2% and 74.9 + 2%, inclusive.

4. Core plane regions within + 2% of core height (+ 2.88 inches) about the bank dednd position of the bank "D" or part length control rods.

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g. With F exceeding F ~ the effects of F on F evaluafEd to determine,if F (Z) is withiNYitsikm(Z)shallbe it.

9 4.2.2.3 When gF (Z) is measured pursuant to specification 4.10.2.2, an overall measurec F (Z) shall be obtained from a power distribution map and increased by 39 to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty.

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Fff I.yg & >< OS Vr~, 2 1 6 4 4 7 1. 6 4 -

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7 FARLEY - UNIT 1 3/4 2-7 l

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i POWER DISTRIBUTION LIMITS BASES M QQ W'b-ORS-3/4.2.2 and 3/4.2.3 HEAT FLUX AND NUCLEAR ENTHALPY HOT CHANNE ',

, Fg (Z):::::3F H u Fgs)

- The limits on heat flux and nuclear enthalpy hot channel factors'

. ensure that 1) the design limits on peak local power density and minimum DNBR are not exceeded and 2) in the event of a LOCA the peak fuel clad f~ temperature will not e.xceed the 2200*F ECCS acceptance criteria limit.

Each of these hot channel factors are measureable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3. This periodic surveillance is sufficient to insure that the hot channel factor limits are maintained provided:

a. Control rods in a' bank move together with no individual rod insertion differing by more than + 12 steps from the group _

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. demand position. ,

b. Control rod banks are sequenced with overlapping banks as described in Specification 3.1.3.5.

c. The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained. .

' d. The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.

N as a function of THERMAL POWER allows changes The relaxation in F for all permissible rod, insertion limits.

igtheradialpowershaphgi its limits provided conditions a through d F wi ab$ve,llbemaintainedwithn are maintained. ,

measurement is taken, both experimental error and man-L" n an Fn 5% is the appropriate allowance ufactur..ig tolMrance must be allowed for.for a full core map taken with the i 3% is the appropriate allowance for manufacturing tolerance.

When F N is measured, experimental error must be allowed for and 4%

istheapprbhriateallowanceforafullcoremaptakenwiththeincorealso g contains an 85 allow-detection system. The specified limit for F,g agte for uncertainties which mean that normat operation will result in .--

F 1 1.55/1.08. ,

tiHons:

FARLEY - UNIT 1 B 3/4 2-4

POWER DISTRIBUTION LIMITS BASES

a. Abnromal perturbations in ghe radial power shape, such as from rod misalignment, effect FaH more directly than Fq,

b. Although rod mnvement has a direct influence upon limiting F to withy.n its limit, such control is not readily available to limit FlH, and r- c. Errors in prediction for control power shape detected dur.ing

( startup physics tests can be compensated for in F n bygestrict-ing axial flux distributions. This compensation Yor F 3g is ,

less readily available.

I 3/4.2.4 QUADRANT POWER TILT RATIO The quardrant power tilt ratio limit assures that the radial power distribution satisfies the design values used in the power capability

1. . - analysis. Radial power distribution measurements,are made during start-up testing and periodically during power operation.

The limit of 1.02 at which corrective action is required provides DNB and linear heat generation rate protection with x-y plane power til ts . A limiting tilt of 1.025 can be tolerated before the margin for uncertainty in Fg is depleted. The limit of 1.02 was selected to provide an allowance for the uncertainty associated with the indicated power tilt.

t The two hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correction of a dropped or n.isaligned rod. In the event such action does not correct the tilt, the margin for uncertainty on Fn is reinstated by reducing the power by 3 percent for each percent of tilt in excess of -

1.0.

k The radial peaking factor, Fxy (z), is measured periodically ro -

to Fg p(z) , vide additional remains withinassurance that Fxy its limit. The the hot (z) channel factor, limits were determined from expected power control maneuvers over the -

full range of burnup conditions in the core. , .

FARLEY - UNIT 1 B 3/4 2-5

SAFETY EVALUATION FOR DELETION OF REACTOR TRIP FOLLOWING TURBINE TRIP BELOW 50% POWER

Background:

The current Farley reactor protection system design provides for a direct reactor trip _following a turbine trip when the plant is above 10-percent power. Since Farley is designed for 50-percent load re-jection capability, a reactor trip following turbine trip below 50%

power can be eliminated without compromising adequate safety margins.

Deletion of the reactor trip following turbine trip would significantly reduce the down time required if the cause of the turbine trip is readily correctable.

References:

r (a) Technical Specification Table 3.3-1 and Page B 2-7.

t (b) FSAR Section 15.2.7 Bases:

An evaluation and analysis has been performed to ensure that the deletion of reactor trip following turbine trip from 50-percent power or less has no adverse affect on plant safety. This evaluation

.. consisted of:

(1) A verification of the worst case transient with respect to core limits in the 10 to 50-percent power range is acceptable (i.e. , minimum DNBR > 1.30),

(2) The consideration of the worst single active failure (unsuccessful reactor coolant pump bus transfer 30 g

seconds after turbine trip), and (3) The acceptability of potential offsite doses resulting from the loss of the condenser and consequential atmospheric -

dumping of steam through the main steam line safety valves.

Results of the analysis show that the plant design is such that.

a turbine trip without a direct or immediate reactor trip from 50 percent power or less presents no hazard to the integrity of the RCS, the main steam system, or the general public. Pressure relieving devices incorporated in the two systems are adequate to limit the maximum pressures to within the design limits. The analysis also demonstrates that for a complete loss of forced reactor coolant flow initiated from the most adverse preconditions ,

of a turbine trip, the DNBR is well above 1.3 at any time during the transient. Thus, no fuel or clad damage is predicted, and all applicable acceptance criteria are met. In addition, the potential offsite doses associated with the loss of the condenser and consequential

Safsty Evaluatioa for Deletion of Reactor Trip Following Page 2 Turbine Trip Below 50% Power (Continued) t atmospheric dumping of steam through the main steam safety valves-were found to M well within the Appendix I limits.

Conclusion:

Deletion of reactor trip following turbine trip below 50% power does not constitute an unreviewed safety question under 10CFR50.59.

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l TABLE 3.3-1 (Continued)

ACTION 9 - With a channel associated with an operating loop inoperable, restore the inoperable channel to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; however, one channel associated with an operating loop may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing per Specification 4.3.1.1.1.

ACTION 10 - With one channel inoperable, restore the inoperable channel to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to below P-8 within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Operation below P-8 may continue pursuant to ACTION 11.

ACTION 11 - With less than the Minimum Number of Channels OPERABLE, r~ operation may continue provided the inoperable channel

(. is placed in the tripped condition within .1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

ACTION 12 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoper:ble channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in HOT STAND 3Y within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and/or open the reactor trip breakers.

REACTOR TRIP SYSTEM INTERLOCKS DESIGNATION CONDITION AND SETPOINT FUNCTION P-6 With 2 of 2 Intermediate Rangej ) P-6 defeats the manual Neutron Flux Channels < 6 x 10 block of source range amp s' . reactor trip.

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's. P-7 With 2 of 4 Power Range Neutron P-7 defeats the automatic Flux Channels > 11% of RATED block of reactor trip THERMAL POWER or 1 of 2 Turbine on: Low flow in more i impulse chamber pressure channels than one primary coolant -

> 55 psia.

loop, reactor coolant pump under-voltage.and under-frequency, ' ^' : l em-p, pressurizer low l pressure, and pressurizer high level .

m FARLEY - UNIT 1 3/4 3-7 3

1 T ABLE 3.3-1 (Continued)

FUNCTION _

CONDITION AND SETPOINT DESIGNATION P-8 defeats the automatic With 2 of 4 Power Range Neutron block of reactor trip P-8 Flux channels > 36% of RATED on low coolant flow in THERMAL POWER. a single loop.

P-10 prevents the manual P-10 With 3 cf 4 Power range neutron block of: Power range flux channels < 8% of RATED low setpoint reactor THERMAL POWER. trip, intermediate j range reactor trip, and i,

intermediate range rod stops.

Provides input to P-7.

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trS- Rea.cf6n Te(f V) Ck A.h h e.Js >_. s/ vo 7"" 3'"' '" EP' s- RATcC> TWERMAL.

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FARLEY - UN]T I 3/4 3-8

LIMITING SAFETY SYSTEf4 SETTIf1G5 BASES reliability of the Reactor Protection System. This trip is redundant to the Steam Generator Water Level Low-Low trip. The Steam /Feedwater Flow Mismatch portion of this trip is activated when the steam flow exceeds the feedwater flow by > 1.55 x 10 6 lbs/ hour. The Steam Generator low Water level portion of the trip is activated when the water level drops below 25 percent, as indicated by the narrow range instrument. These trip values include sufficient allowance in excess of normal operating values to preclude spurious . trips but will initiate a reactor trip before the steam generators are dry. Therefore, the required capacity

( and starting time requirements of the auxiliary feedwater pumps are i reduced and the resulting thermal transient on the Reactor Cooiant System and steam generators is minimized.

Undervoltage and Underfrequency - Reactor Coolant Pump Busses The Undervoltage and Underfrequency Reactor Coolant Pump bus trips provide reactor core protection against Of48 as a result of loss of The voltage or underfrequency to m, ore than one reactor coolant pump.

specified set points assure c reactor trip signal is generated before the low flow trip set point is reached. Time delays are incorporated in the underfrequency and undervoltage trips to prevent spurious reactor trips from momentary electrical power transients. For under-voltage, the delay is set so that the time required for a signal to reach the reactor trip breakers following the simultaneous trip of two or more reactor ~ coolant pump bus circuit breakers shall not exceed 0.9

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seconds. For underfrequency, the delay is set so that the time required for a' signal to reach the reactor trip breakers after, the underfrequency trip set point is reached shall not exceed 0.3 seconds.

Turbine Trip A Turbine Trip causes a direct reactor trip when operating above P- .

Each of the turbine trips , provide turbine protection and reduce tte severity of the ensuing transient. fio credit was taken in the accident analyses for operation of these trips. Their functional capability at the specified trip settings is required to enhance the overall reliability of the Reactor Protection System.

b FARLEY - UtilT 1 B 2-7

. - - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ . . . _ . _ _. _ _ _ \

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SAFETY EVALUATION FOR REMOVAL OF THE PAP,T LENGTH CONTROL. RODS Backaround:

Part length control reds are presently instai'cd in the Farley-Unit 1 reactor vessel . Utilization of the part length rods is currently prohibited during power operation by the technical specifications. Removal of the part length rods will decrease outage time and radiation exposure associated with surveillance testing. For these reasons, Alabama Power Company plans to remove the part length control rods during the first refueling outage. Thus, it will be necessary to amend the technical specifications to re-flect the deletion of the part length control rods prior to startup folicwing the first refueling outage.

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References:

(' (a) Technical Specifications 1.13, 3.1.3.1, 4.1.3.1.2, Tabl e 3.1-1, 3.1.3.2, 3.1.3.3, 3.1.3.7, 4.1.3.7, 4.2.2.2, 3.2.4, 3.10.1, 4.10.1.3, 3.10.2, 3.10.3,B 3/4.2.1 and 5.3.2.

(b) FSAR Sections 4.2.3 and 15.0.

Bases:

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The part length control rods are presently excluded from use during plant operation per the technical specifications to assure that the axial power shape used in the safety analyses in the FSAR is met. ,

The part length control rods will be removed during the first '

refueling outage. Thimble plug assemblies will be installed

(' into the locations previously occupied by the part length rods.

l l ,_ These plugs are installed to preserve the current dynamic operat-ing characteristics of the reactor vessel, i.e. pressure drops, coolant flow rates, etc. The following factors were considered regarding the installation of the thimble plugs.

(1) Fuel assemblies without control rods, burnable poison rods, or source rods use identical de-j vices.

(2) The thimble plugs will effectively limit by-f 4 pass flow just as they currently limit bypass I flow in those assemblies which do not contain control rods, source rods, or burnable poison -

rods.

l (3) Since the plugged fuel assemblies have no ad-verse effect on the design core flow distribu-tion, the calculated core thermal margin will be unaffected.

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(4) Physics analyses indicate that there will be no adverse effect by the plugged assemblies on the physics characteristics of the core.

(5) No information exists which suggests that the replacement of the part length control rods with thimble plug assemblies increases the probability of any accident previously analyzed.

(6) No information exists which suggests that the replacement of the part length control rods with thimble plug assemblies introduces a possibility for an accident or any malfunction of a different type than those previously

- analyzed.

- (7) Permanent removal of the part length control rods will further minimize the probability '

of an event which could cause an abnormal power distribution.

Thus, removal of the part length control rods will not reduce the margin in the safety analyses or create an accident not previously analyzed.

Conclusion:

Removal of the part length control rods does not constitute l

, an unreviewed safety question as defined by 10CFR50.59.

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INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION Paoe 3/4.0 APPLICABILITY............................................. 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 B0 RATION CONTROL .

Shutdown Margin - T > 200*F......................... 3/41-1

.. av9 Shutdown Margin - T,yg 1'200'F......................... 3/4 1-3

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B o ro n Di 1 u t i o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1 4 Moderator Tempe ra ture Coef fi cient. . . . . . . . . . . . . . . . . . . . . . 3/4 1-5 Mi nimum Tempera ture for Cri ti cal i ty. . . . . . . . . . . . . . . . . . . . 3/4 1-6 3/4.1.2 B0 RATION SYSTEMS Fl ow Pa t h s - S h u td own . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-7 m

A,_) Flow Paths r Operating................................. 3/4 1-8 Charging Pump - Shutdown............................... 3/4 1-9 Charging Pumps - Operating............................. 3/4 1-10 Boric Acid Transfer Pumps - Shutdown...................

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3/4 1 -11 Boric Acid Transfer Pumps - 0perating. . . . . . . . . . .. . . . . . . 3/4 1-12 ,

l Ba rated Wa ter Sources - Shutdown. . . . . . . . . . . . . . . . . . . . . . . 3/4 1-13 Bora ted Wa ter Sources - Operating. . . . . . . . . . . . . . . . . . . . . . 3/4 1-14 3/4.1.3 MOVABLE CONTROL ASSEMBLIES Group Height........................................... 3/4 1-16 Posi tion Indica tor Channels Operating . . . . . . . .. . . . . . . . . 3/4 1-19 Posi tion Indicator Channel s Shutdown. . . . . . . . . . . . . . . . . . . 3/4 1-20 Rod Drop Time.......................................... 3/4 1-21 Shutdown Rod In serti on Limi t. . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-22 Control Rod Ins erti on Limi ts . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-23

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pf.!jNEL FU::cTIONAL TEST 1.11 A CHf.NNEL FUNCTIONAL TEST shall be the injection of a simulated signal into the channel as close to the primary senscr as practicable to verify OPERABILITY including alarm and/or trip functions.

C03E ALTfRATION 1.12 CORE ALTERATIO.:' ahall be the povement or manipulation of any com-ponent within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of CORE At_TERATION shall not preclude completion of movement of a component to a safe conservative position.

SHUTCO'.:M 14ARGIri 1.13 SHUTDOWi ilARGlti shall be the instantaneous amount of reactivity by which the reactor is subtritical or would be subcritical from its pre-sent condition assumingj;)

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l bT, b,- All full length rod cluster assemblies (shutdown and control)

's ___ are fully inserted except for the single rod cluster assembly s.

of highest reactivity worth which is assumed to be fully withdrawn. .

IDEtTIFIED LEAKAGE 1.14 IDENTIFIED LEAKAGE shall be:

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() a. Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are etptured and conducted to a sump or collecting tank, or

b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere l

with the operation of Icakage detection systems or not to be' PRESSURE BOUNDARY LEAKAGE, or

c. Reactor coolant system leakage through a steam generator to the secondary system.

UtilDENTIFIED LEAKAGE 1.15 UllIDEllTIFIED LEAKAGE shall be all leakage which is not IDEllTIFIED LEAKAGE or CONTROLLED LEAKAGE.

FARLEY - UtilT 1 1-3

O REACTIVITY CONTROL SYSTEMS 3/4.1.3 MOVABLE CONTROL ASSEMBLIES _

GROUP HEIGHT LIMITING CONDITION FOR OPERATION I

3.1.3.1 All full length (shutdown and control) rods :M :" p.-t 1x;th eede which are inserted in the core shall be OPERABLE and positioned within + 12 steps (indicated position) of their group step counter

,]demandposition.

APPLICABILITY: MODES 1* and 2*

ACTION:

a. With one or more full length rods inoperable due to being immovable as a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and be in HOT STANDBY within @

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

t

b. With more than or.e full length rod inoperable or misaligned from the g: cup step counter demand position by more than + 12 steps (indicated position), be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

c. With one full. length rod inoperable or misaligned from its ,

(' group step counter demand position by more than + 12 steps (indicated position), POWER OPERATION may continue provided that within one hour either:

1. The rod is restored to OPERABLE status within the above alignment requirements, or

2. The rod is declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied.

' POWER OPERATION may then continue provided that:

a) A revaluation of each accident analysis of Table l 3.1-1 is performed within 5 days, this reevaluation shall confirm that the previously analyzed results .

of these accidents remain valid for the duration of

/ operation under these conditions, l

b)

The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, g and

• See Special Test Exceptions 3.10.2 and 3.10.3.

FARLEY - UNIT 1 3/4 1-16

REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION (Continued) c)

A power incore distribution detectors and map F (Z) and Fis obtaingd from the movable within their limits wikhin 72 houN.are verified to be d) Either the THERMAL POWEP level is reduced to < 75%

of RATED THERMAL POWER within one hour and within the

. next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> the high neutron flux trip setpoint is reduced to < 85% of RATED THERMAL POWER, or

(-

( e) The remainder of the rods in the group with the inoperable rod are aligned to within + 12 steps of the inoperable rod within one hour while Eiaintaining the rod sequence and insertion limits of Figures 3.1-1 and 3.1-2; the THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation.

SURVEILLANCE P.EQUIREMENTS ,

(v 4.1.3.1.1 The position of each full length rod shall be determined to be within the group demand limit by verifying the individual rod positions at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except during time intervals when the Rod Position Deviation Monitor is inoperable, then verify the group positions at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

4.1. 3.1. 2 Each full length rod not fully inserted cr> ::.d ;:rt 1-..p;r l

' ' in the core shall be determined to be OPERABLE by l movement of at least 10 steps in any one direction at least once per 31 days.

O_

FARLEY - UNIT 1 3/4 1-17

P O

TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN IN0PERABLE FULL M LENGTH R00 Rod Cluster Control Assembly Insertion Characteristics Rod Cluster Control Assembly Misalignment loss Of Reactor Coolant From Small Ruptured Pipes Or From Cracks In Large Pipes Which Actuates The Emergency Core Cooling System Single Rod Cluster Control Assembly Withdrawal At Full Power Major Reactor Coolant System Pipe Ruptures- (Loss Of. Coolant Accident) h Major Secondary System Pipe Rupture Rupture of a Control Rod Drive Mechanism Housing (Rod Cluster Control Assembly Ejection)

~

f

~

a FARLEY - UNIT 1 3/4 1-18

REACTIVITY'CONTROI SYSTEMS

.Q DOSITION INDICATOR CHANNELS OPERATING _

LIMITING C0!:DITION FOR OPERATION l

3.1.3.2 All shutdown, :r- t. ;' and ;

_--;t' control rod position indi-cator channels and the demand position indication system shall be OPERABLE l and capable of determining the control rod positions within + 12 steps.

APPLICABILITY: MODES 1 and 2.

ACTION:

a. With a maximum of one rod position indicator channel per group

( inoperable either:

1. Detennine the position of the non-indicating rod (s) in-directly by the movable incore detectors at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and innediately af.ter any motion of the non-indicating rod which exceeds 24 steps in one direction since the last determination of the rod's position, or

2. Reduce THERMAL "0WER TO < 50% of RATED THERMAL POWER within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

O b. With a maximum of one demand position indicator per bank inoperable either:

1. Verify that all rod position indicators for the affected bank are OPERABLE and that the most withdrawn rod and the least withdrawn rod of the bank are within a maximum of -

i 12 steps of each other at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, or

2. Reduce THERMAL POWER to < 50% of RATED THERMAL POWER within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

c. The provision of Specification 3.0.4 are not applicable.

SURVEILLANCE REOUIREMENTS 4.1.3.2 Each rod position indicator channel shall be determined to be OPERABLE by verifying the demand position indication system and the rod position indicator channels agree within 12 steps at least once per 12 -

hours except during time intervals when the Rod Position Deviation Monitor is inoperable, then compare the demand position indication system and the rod position indicator channels at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

( FARLEY - UNIT 1 3/4 1-19 res :

- t -, ,

5-I

C REACTIVITY CONTROL SYSTEMS POSITION INDICATOR CHANNELS - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.3.3 At least one rod position indicator channel (excluding demand q position indication) shall' be OPERABLE for each shutdowny' control s>F- r o cl 4e -t larct' ri not fully inserted.

[ iPPLICABILITY: P.0DE5 3,* 4* and 5*.

ACTION: With less than the above required position indicator channels OPERABLE, immediately open the reactor trip system breakers.

~

n SURVEILLANCE REQUIREMENTS 4.1.3.3 Each of the abo /e required rod position indicator channels shall

-- be determined to be OPERAb E by verifying the demand position indication system and the rod position indicator channels agree within 12 steps at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

l l

i

• With the reactor trip system breakers in the closed position.

I

1. ^

l l -

~

FARLEY - UNIT 1 3/4 1-20 l . . . . _ __ __

e ,+--

- - ~ -u ,

REACTIV1TY C0t1 TROL SYSTEMS PART LEtiG R0D If1SERTIO!1 LIMITS LIMITIf1G C0tiDI 014 FOR OPERATI0fi 3.1.3.7 All part i ngth rods shall be fully withdrawn.

APPLICABILITY: MODES

• and 2*

ACTI0ti:

With a maximum of one part ngth rod not fully ithdrawn, within one hour either:

a. Fully withdraw the rc or

b. Be in HOT STAriDEY withi he n xt 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

-SURVEILLAf1CE REOUIREMEtiTS x 4.1. 3. 7 Each part length rod hall be determi d to be fully withdrawn by:

/ a. Verifying the po tion of the part leng rod prior to increasing THERMAL POWER 'ove 5% of RATED THERMAL POWER, and

b. Veri fying, least once per 31 days, the ctric power has been disconnecte from its drive mechanism by physkal removal of a breaker 6r the circuit.

See Specia est Exception 3.10.2 and 3.10.3.

k FARLEY - UtilT 1 3/4 1-26

~ - .

(. P0"iin DISTPliGil0i! Lil'ITS SUPVFIll Auf L REOUIREMD:f 5 {Continu :,')

~

2. When the F C

is less than or equal to the F RTP linit for the appropN, ate neasured core plane, adjitiUEnl power dfr+,ributiop g naps siali he tchn and F coi.: pared to F

yy and F xy at least once per 31 EFPDYY

e. The F limits for RATED TiiER"AL FO', ER uithin specific core plane [Yshall be:

RTP

l. F 1 1.71 for all core planes ccataining bank "D" 1

(

cdh, trol rods = - ~'f- . . .. , , . . .. r A , and RTP

2. F xy 1 1.55 for all unrodded core planes.

f. The F limits of e, above, are not applicable in the fol-lowinhYcore plane regions as measured in percent of core height from the bottom of the fuel:

1. Lower core region from 0 to 15%, inclusive.

~

2. Upper core region from 85 to 100%, inclusive.

3. Grid plane regions at 17. 8 + 21' , 32.1 + 2 % , 46. 4 + 20.' ,

60.6 + 2% and 74.9 + 2%, inclusive. codwl **b

4. Core plane regions t;ithin + 2% of core height (+ 2.83 [

inches) about the bank demind position of the bink "D",e- 1 p- , c,s,, -

9 With F exceeding F the effects of F on Fg (Z) shall be evaluafEJtodetermiEE,ifF(Z)iswithiEYits q 1Tmit.

l 4.2.2.3 Uhen F (Z) is measured pursuant to specification 4.10.2. 2, an overall measure F and increased by 39(Z) shall be to account forobtained nanufacturing from tolerances a power distribution and furthermap increased by SZ to account for measurement uncertainty.

1 FARLEY - UillT 1 3/4 2-7 ,

1

() PO'.-!EP DISTRIBUT1071 1.lMITS gufDRAilf POWER TILT RATIO LIMITil:G COWL;fl0N TOR OPERAT10:1 3.2.4 Tl!E QUADRAf1T POUER TILT RATIO shall not exceed 1.02.

APPLICAD!LITY: i;0DE 1 AB0VE 50% CF RATED ll!ERMAL POWER

• ACTIO:i:

c.

(/ , With the QUADRAllT POWER TILT RATIO determined to exceed 1.02 but < l.09:

1. Uithin 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:

a) Either reduce the QUADRAftT PCUER TILT RATIO to.

uithin its limit, or b) Reduce TiiEREL POWER at least 3R from RATED TilERMAL PC'aR for each 1% of indicated QUADRAllT POWER TILT RATIO in excess of 1.0 and similarly reduce the c--l . Po'.rer Range fleutron Flux-High Trip Setpoints within i the ncxt 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

2. Verify that the QUADRAl!T POWER TILT RATIO is within its limit-within E4 hours after exceeding the limit or reduce TIIERMAL PDWER to less than 50% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and reduce the' Power Range Heutron Flux-ilich Trip set-

-!- points to 1 5% 5 of RATED TilERML POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

3. Identify and correct the cause of the out of limit condition prior to increasing' THERMAL POWER; subsequent POWER OPERATION above 50;; of RATED l}lERMAL power r:ay proceed provided that the QUADRAMT POWER TILT RATIO is verified within its limit at least.

once per hour for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or until verified acceptable at' 95?'

or greater RATED T}iERMAL POWER.

b. With the QUADRAtlT POWER TILT RATIO determined to exceed 1.09 due to misalignment of either a shutdown [controlTr. sue-  !

hwy =h rod: g

l. Reduce Tl!ERt:AL POWER at least 37 from RATED TilERMAL ,

POWEP. fer each 1% of indit.ated QUADRANT POWER TILT RATIO in excess of 1.0, within 30 minutes.

2. Verify that the QUADRA!1T POWER TILT RATIO is within its limit

,, within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after exceeding the limit or 5

• See Special Test Exception 3.10.2.

FARLEY - UNIT I 3/4 2-11

t 3/4.10 SPECIAL TEST EXCEPTIONS SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.10.1 The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 may be suspended for measurement of control rod worth and shutdown margin providedi

p. 2[eactivity equivalent to at least the highest estimated f . control rod worth is available for trip insertion from OPERABLE control rod (s), and

(/ _

m. ',;g_.. _ i :t = it" ~ +n =+ h r+ tS 120 :tcp g n : : .. a..e Ondi:.

L APPLICABILITY: MODE 2.

ACTION:

g a. With any full length control rod not fully inserted and with less than the above reactivity equivalent available for trip insertion e J.; }

g . 1 ; c.;; t ' m d e ** wi +M - th:i ...M.jc;.21 1-:-it , initiate and continue boration at > 30 gpm of between 7000 and 7700 ppm boric acid J solution or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.

b. With all full length control rods inserted and the reactor subc'itical r l

{ (K l.0) by less than the above reactivity equivalent, immediately inTba<te and continue boration at > 30 gpm of between 7000 and 7700 ppm boric acid solution or its equivalent until the SHUTDOWN l

MARGIN required by Specification 3.1.1.1 is restored.

l SURVEILLANCE REQUIREMENTS 1

4.10.1.1 The position of each full length rod either partially or fully withdrawn shall be determined ~ at least once per 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

4.10.1.2 Each full length rod not fully inserted shall be demonstrated capable of full insertions when tripped from at least the 50% withdrawn position within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to reducing the SHUTDOWN MARGIN to less than the limits of Specification 3.1.1.1.

s.;C.l.: N nart length rods if enorni':2 Aui s ce demonstrated ~

q OPERABLE by moving eeu. t ' % d > 10 steps within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> P prior to red c ' . t. HUTDOWN MARGIN to less tnoo o. 1 " t" n (

l 5a ~ ion 3.1.1.1. '

i FARLEY-UNIT 1 3/4 10-1

(P SPECIAL TEST EXCEPTIONS GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION a

3.10.2 The group height, insertion and power distribution limits of Specifications 3.1.3.1, 3.1. 3. 5, 3.1. 3. 6, 0. ' . 3. ? , 3.2.1, and l 3.2.4 may be suspended during the performance of PHYSICS TESTS provided:

The THERMAL POWER is maintained < 85% of RATED THERMAL POWER, a.

and

( b. The limits of Specifications 3.2.2 and 3.2.3 are maintained and determined at the frequencies specified in Specification 4.10.2.2 below. .

APPLICABILITY: MODE 1 1

4 ACTION:

__ q With any of _the ii. rats of Specifications 3.2.2 or 3.2.3 being exceeded W while the reanirements of Specifications 3.1. 3.1, 3.1. 3. 5, 3.1. 3. 6, 0. ' . : . ', {

3.2.1 and 3.2.4 are suspended, either:

a. Reduce THERMAL POWER sufficient to satisfy the ACTION require-ments of Specifications 3.2.2 and 3.2.3, or ,

s b. Be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.10.2.1 The THERMAL POWER shall be determined to be < 85% of RATED THERMAL POWER at least once per hour during PHYSICS TEXTS.

4.10.2.2 The Surveillance Requirements of Specifications 4.2.2 and 4.2.3 shall be performed at the following frequencies during PHYSICS TESTS.

a. Specification 4.2.2 - At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b. Specification 4.2.3 - At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

-g

.FARLEY-UNIT 1 3/4 10-2

?. .

O SPECIAL TEST EXCEPTIONS PHYSICS TESTS LIMITING CONDITION FOR OPERATION 3.10.3 The limitations of Specifications 3.1.1. 4, 3.1. 3.1, 3.1. 3. 5 3.1.3.6_ ---f . .;.~ .nay be suspended during the performance of  :

PHYSICS TESTS provided:

a. The THERMAL POWER does not exceed 5% of PATED THERMAL POWER, and

!. b. The reactor trip setpoints on the OPERABLE Intermediate and Power Range Nuclear Channels are set at < 25% of RATED THERMAL POWER.

APPLICABILITY: MODE 2.

ACTION:

With the THERMAL POWER > 5% of RATED THERMAL POWER, immediately open the reactor trip breakers.

h_

SURVEILLANCE REOUIREMENTS 4.10.3.1 The THERMAL POWER shall be determined to be < 5% of RATED ~

THERIGL POWER at least once per hour during PHYSICS TEXTS.

(

4.10.4.2 Each Intermediate and Power Range Channel shall be subjected to a CHANNEL FUNCTIONAL TEST within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> prior to initiating PHYSICS TESTS.

O FARLEY-UNIT 1 ,

3/4 10-3 m--

(

DESIGN FEATURES DESIGN PRESSURE AND TEMPERATURE 5.2.2 The reactor containment building is designed and shall be main-tained for a maximum internal pressure of 54 psig and a temperature of 280 F.

5.3 REACTOR CORE l'

FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 157 fuel assemblies with each fuel assembly containing 264 fuel rods clad with Zircaloy -4. Each fuel rod shall have a nominal active fuel length of 144 inches and '

contain a maximum total weight of 1766 grams uranium. The initial core loading shall have a maximum enrichment of 3.2 weight percent U-235.

Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrichment of 3.5 weight percent U-235. m (- -

CONTROL R0D ASSEM5 LIES _

5.3.2 The reactor core shall contain 48 full length ead : p-. t I 2 2 : control rod assemblies. The full length control rod assemblies Th; p;ct shall contain a nominal 142 inches of absorber material.

.._1

: . . J . _ ; ;?

i ': ;'h c " r ' --  ::___1._; .; il-c;..t;.

1... . :nJ:. The nominal values of absorber material shall be 80 percent silver,15 percent indium and 5 percent cadmium. All control rods shall be clad with stainless steel tubing. 1 Ik bh:2 :' tt; . .J 1:q .. m U. ,.; c t L. 3... .4 ' m t: F

[

al. - mvid:_

5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained:

(

1 FARLEY - UNIT 1 5-4 1

t 3/4.2 P0i!2R DISMIBUTI0.1 LI"ITS

' BE5ES ___. __.- _ _ _ _ - . . _ . _ _ _

The specifications of this section provide assurance of fuel intcg-rity during Condition I (l'ormal Op; ration) and II (Incidents of I:cderete Fiequency). events by: (a) maintci;iing the minit.ua C?!3R in the core > l.30 _

during norr.al operation and in short tern transients, and (b) limiting .

the fission gas release, fuel pellet temperature & cladding techanical properties to within ast.umed design criteria. In addition, limitin';.the paak line:'r power density during Conditior I events provides assurance 7

that the initial conditions assuusd for the 1.0CA analyses are net and the ECCS acceptance ciiteria limit of 2200 F is not exceeded.

The definitions of certain hot channel and peaking factors as used in these specifications are as follows:

F0 (2) Heat Flux Hot Channel Factor, is defined as the naximum local heat flux en the surface of a fuel rod at core elevation Z divided by the ' overage fuel rod heat flux, allowing the man-ufacturing tolerances on fuel pellets and rods.

._ p il f:aclear Enthclpy Rise llot Channel Factor, is defined as the Idi ratio of the integral of linear power along the rod with the highest integrated power to the average rod pouar.

F Radial Peaking Factor, is defined as the ratio of peak power XY(Z) density to average power density in the horizontal plant at

( core elevation Z.

3/4.2.1 AXIAL FLUX DIFFERENCE (AFD),

The limits on AXIAL FLUX DIFFEREi!CE assure that the Fg (Z) upper bound envelope of 2.32 times the normalized axial peaking ractor is not .

exceeded during either normal operation or in the event of xenon redis-tribution following power changes.

Target flux difference is determined at equilibrium xenon conditions.

The full

' ~

' ' ' + '

_.'.m.- . ,u.-.. . .. .. -  ;

length rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal position for steady state operation at high power. levels. The value -

of the target flux difference obtained under these conditions devided by the fraction of RATED THERilAL POWER is the target flux difference at RATED THER!iAL POWER for the associated core burnup conditions.

Target flux dif ferences for other TilEG;AL PO'.lER Icvels are obtained by multiplying the RATED TilERMAl. PO'iER value by the appropriate fractional difference value is necessary to reflect core burnup considerations.

FARI.EY - Ui!IT 1 B 3/4 2-1

/

. f * ..

(!

PG'.!T *: IriST!:IBU110;; 1.I!-lITS MSES .. . . _ _ . _ . . . _ _ _ . _ _ _ _

I.lthough it is intended that the plant will be opert.ted with ti,e A7.lAt FLUT. DIFFERIiX.E uithin the f.(5)? teruet tand tbaut the target flu /.

deffererice, during rcpid plcat THUD.L PO'.ER reductions, control rod it.otion will cause the AFD to davicte outside of the target band at re-duced T:!'iR:D._ P01:iR icvels. This deviatien will not affect the xenon redistribution sufficiently to chN.ga the envelopa of peak.ing factors

( ubich rny be reached on a subsee,urnt return to RATED Tile!CML PC'!ER (uith the AFD within th2 target band) providsd the tima duration of the devi-ation is limited. l.ccordingly, a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> penalty deviation limit cumulation during the pre /ious 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is provided for operation outside of the target band but within the limits of Figure 3.2-1 while at Tl!ER".AL POWER levels beteetn 50% and S05 of RATED TilEFJ1AL PC' DER. From THEPPAL P0i!ER levels bett:ren ISS and 50% of LATED THEIJiAL PO'.!ER, deviations of the ISD outside of ti.e tcrget b:.nd cre less significant. The penalty of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> acutal tir.:2 reflects this reduc,cd significance.

Provisions for monitoring the AFD cn en automatic basis are derived from the plant process combuter though the AFD Monitor Alara. The computer ^'C-determines the ona minute everage of each of the OPERABLE excore detector outputs and provides en clara nassage it.ediately if the AFD for at least 2 of 4 or 2 of 3 OPERABLE cy. core' channels are outside the target band and the THEP.:iAL P0'IER is gretter than 90% of RATED TifEfJ:AL PC'lER. During operation at THER"AL PONER levels between 50% and 90% between 15% and 50r

\ RATED Ti!EEML POWER, the computer outputs an alarm massaoe with the penalty deviation accumulates beyond the limits of I hour and 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, respectively.

Figure B 3/4 2-1 st(ows a typical nonthly target band.

/

I

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._ o Faut.LY - UNIT 1 B 3/4 2-2

i SAFETY EVALUATION FOR REVISION TO TECHNICAL SPECIFICATION FIGURE 6.2-2 AND TECHNICAL SPECIFICATIONS 6.5.1.2 AND 6.5.3.1 Recommended Change:

See attached sheets.

Background aci Bases:

The proposed change revises section 6.0, Administrative Controls, of the Joseph M. Farley Unit 1 Technical Specifications to reflect recent changes to the facility organization (i.e. the addition of Maintenance Superintendent to assume responsibility for Maintenance-and Instrumentation and Control activities). The addition of the r-l Maintenance Superintendent to the PORC as a voting member will not affect the ability of the PORC to perform its function, that of advising the Plant Manager os all matters related to nuclear safety.

The Maintenance Superintendent will add additional expertise to the PORC.

s The division of Opeest i.ons, Maintenance and Instrumentation and Control activities under two Superintendents increases the effec-tiveness of the review and approval process for procedures, and changes thereto, which affect plant nuclear safety.

Conclusion:

The proposal changes do not involve an unreviewed safety question as defined by 10CFR50.59.

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MANAGEM ,

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ADMINISTRATIVE CCNTROLS 6.3 FACILITY STAFF OUALIFICATIONS 6.3.1 Each member of the faci.lity staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for ccmaarable positions. except for the Chemistry and Health Physics Supervisor who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975.*

6.4 TRAINING ,

6.a.1 A retraining and replacement training program for the facility staff shall be maintained under the direction of the Training Supervisor and shall meet or exceed the requirements and reccmendations of Section'

(

i 5.5 of ANSI N18.1-1971 and Appendix "A' of 10 CFR Part 55.

6.4.2 A training program for the Fire Brigade shall be maintained under the direction of the Training Suoervisor and shall meet or exceed the requirements of Section 27 of the NFPA Code-1976, except for Fire Brigade training sessions which shall be held at least quarterly.

5.5 REVIEW AND AUDIT 6.5.1 PLANT OPERATIONS REVIEW COMMITTEE (PORCl

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FUNCTICN 6.5.1.1 The PCRC shall function to advise the Plant Manager on all matters related to nuclear safety.

,. COM.FOSITIC.t

<N 6.5.1.2 The PORC snall be composed of the: .

Chairman: Plant Manager Vice Chairman: Assistant Plant Manager

! Member: Technical Suoerintendent Member: Operations Superintendent Member: (Non-Voting) Plant Quality Assurance Engineer 1

trever.r: h oigteg g g ,,i M y ALTERNATES l

S.6.1.3 All alternate members shall be ac:ointed in writing by the FORC Chairman tc serve on a temocrary basis; hcwever, no more than one alternate shall particioate as voting members in PCRC activities at any one time.

Ge Minimum cualifications recuirement for the Chemistry and Health Pnysics Sucervisor shall beccme effective wnen the initial incumoent

( in this positien is replaced.

FARELY - UNIT 1 6-5 Amendment No.

l

[ L1

( ADMINISTRATIVE CONTROLS

a. Minutes of each NORB meeting shall be prepared, approved and forwarded to the Senior Vice President within 14 days following each meeting. ,

b. Reports of reviews encompassed by Section 6.5.2.7 above, shall be prepared, approved and forwarded to the Senior Vice President within 14 days following completion of the review.

c. Audit reports encompassed by Section 6.5.2.8 above, shall be

- forwarded to the Senior Vice President and to the management positions responsible for the areas audited within 30, days after completion of the audit. .

6.5.3 TECHNICAL p.EVIE',I AND CONTROL ACTIVITIES 5 . 5. 3 .1 Activities which affect nuclear safety shall be conducted as follows:

a. Procedures required by Technical Specification 6.8 and other procedures which affect plant nuclear safety, and changes (other than editorial or typographical changes) thereto, shall be prepared, reviewed and aporoved. Each such procedure or procedure change shall be reviewed by an individual / group other

- than the individual / group which prepared the procedure or pro-cedure change, but who may be frem the same organization as the individual / group which are::ared the procedure or procedure

' change. Procedures other than Administrative Procedures will

[ be approved by either the Technical Superintendent, the Opera-

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wYku- nte-- tions Superintendengor the Assistant Plant Manager as applic-

+Spon b b,d nde d a bl e . The Plant Manager will aporove administrative crocedures,

' security implementing procedures and emergency plan implementing procedures. Temporary changes to procedures which clearly do not change the intent of the approved procedures will be accroved by two members of the plant staff, at least one of whom holds a Senior Reactor Operator's License. For changes to procedures which may involve a change in intent of the approved procedures, the person authori:ed above to aporove the procedure shall ap-prove the change.

b. Proposed changes or modifications to plant nuclear safety-related structures, systems and ccaponents snall be reviewed as designated by the Plant Manager. Each such modification ,

shall be reviewed by an individual / group other than the individual / group which designed the nodification, but who r.ay be from the same organi:stion as the individual / group whicn designed the modifications. Proposed modifications to olant

( nuclear safety-related structures, systems and components shall be approved prior to implementation by the Plant Manager.

FARLEY - UNIT 1 6-11 Amendment No. 4' l

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